This invention relates to a device for local measurement of the quantity of power within the fuel of a reactor core fuel assembly. By making use of suitable means, the determination of this quantity at several axial positions in a fuel assembly and in several fuel assemblies serve to ensure reliable and efficient protection of fuel assemblies, therefore of the reactor core and of the entire nuclear power plant.
Many devices for the measurement of local power are already known. In general they are based on the principle of determination of the nuetron flux level in the immediate vicinity of the fuel and a known relationship is used for relating the value of this flux to the power output. Devices of this type which are particularly worth of mention are fission chambers, neutron detectors, neutron thermometers and collectrons. There is not a single instance, however, in which these instruments perform their intended function in a wholly satisfactory manner. The reason for this is that the signals delivered by the instruments are not usually proportional to the local power released by the fuel since they vary with the incident neutron flux. Determination of the power output consequently entails the need for cetain approximations in order to take into account both the reduction in cross-section of spent fuel and the reduction in strength of the signal emitter itself during the period of utilization of the instrument.
It has also been proposed to carry out local measurement of power output by means of instruments known as "gamma-ray thermometers" in which an increase in the temperature under measurement is brought about by exposure to electromagnetic radiation in the form of gamma rays, the proportion of which can attain 95%. Of these latter, approximately 70% result from fissions which have taken place during a time interval of the order of five minutes prior to measurement. In gamma-ray thermometers, the value of the temperature rise in a radiation-absorbing body can in fact be measured along a controlled and constant path provided for the removal of the heat produced. The laws of conduction accordingly show that the temperature difference measured on a given path of constant conductance through the body is proportional to the production of heat and consequently to the power released by the nuclear fuel located in its immediate vicinity as a result of fissions within said fuel.
In actual practice, however, gamma-ray thermometers have been employed up to the present time almost exclusively in heavy-water reactors. Their intended function in these reactors is to deliver signals which are proportional to the specific heat production without requiring any compensation for uranium depletion as the reactor operation proceeds or for loss of sensitivity of the measuring instrument. Moreover, these instruments have very high stability since gamma-ray absorption is dependent solely on the density of the radiation-absorbing body and is not affected by variations arising from modifications of the atomic or isotopic structure which characterize other types of detector. In this case the effect produced by the neutron flux on the material has a negligible consequence in regard to the properties which determine the absorption of gamma rays and the resultant temperature rise.
However, gamma-ray thermometers of the type which have been designed up to the present time, especially for heavy-water reactors, are adapted only to point-measurement powers expressed in mW/g. As a general rule, these instruments make use of an absorbent metallic mass mounted within a protective sheath, the sheath being placed among the fuel elements within the reactor core. A first portion of the metallic mass is in contact with the sheath and the external environment and is substantially at the temperature of this latter while a second portion is connected to the first but located within an insulated chamber, the space between the second portion and the sheath being filled with a gas or air or even evacuated. The device thus forms a heat sink in which the temperature difference between the two portions of the absorbent mass can be measured by means of thermocouples. After calibration, the absorbed thermal power and therefore the power released by the surrounding nuclear fuel can be deduced from a knowledge of the geometrical characteristics of the absorbent mass. However, gamma thermometers up to the present time have been calibrated by only indirect means. For example, it is known that in one application the gamma thermometers are cross-calibrated with neutron flux measurement at the same location. In another situation, gamma thermometers have been calibrated by measurement of transient thermal response outside of the nuclear reactor. It is presently unknown to calibrate either in situ or ex-situ of the reactor by utilizing the Joule effect of applied electrical current.
One of the objects of the present invention is to permit the practical construction of a gamma-ray thermometer array for use in gamma fluxes of appreciably higher value than those encountered in heavy-water reactors. Such a system can be employed, for example, in light-water reactors and even in fast reactors while permitting measurement of the linear power released by the fuel which is usually measured in W/cm.
A further object of the invention is to permit a sufficient degree of miniaturization of the gamma-ray thermometer array under consideration by virtue of the substantially higher rates of temperature rise in a light-water reactor. Thus the system can be mounted within a sheath of very small diameter and of substantial length. Said sheath can be introduced among the bundle of canned fuel pins of a fuel assembly which is employed in the usual manner in a reactor of this type (fuel pins each having a diameter between 8 and 10 mm and a length of 4 m).
Yet another object of the invention is to permit accurate and reliable measurement which is not subject to any appreciable drift in the course of time and is therefore particularly reliable; this measurement can be performed in a series of separate and distinct zones disposed successively in the vertical direction along the entire length of the fuel assembly.
Still another object of the invention is to permit the construction of a gamma-ray thermometer array which is adapted to undergo accurate calibration prior to mounting within the reactor core. Such calibration being carried out by means of a system would produce heat generation that would simulate the power that would be produced in the fuel which simulates the power released by the fuel.
One more object of the invention is to permit mechanical tuning i.e. adjustment, of the gamma ray thermometer system remote from the reactor. Such mechanical tuning being carried out by means which fix the calibration to be within pre-selected desired standards.
As mentioned earlier, the principle adopted in order to carry out the measurement by means of a gamma-ray thermometer consists in very general terms in establishing a path having a constant and predetermined value of heat conductance in the mass of a gamma-absorbing body and, after calibration, in deducing from the temperature difference recorded at the ends of said path the value of the heat which is produced in the body by absorption of the gamma radiation released by the power production in the surrounding nuclear fuel and therefore produced by the surrounding nuclear fuel.
With these objectives, consideration has been given to a number of configurations defining a specific path for the flow of heat produced, reducing the effectiveness of or by eliminating the secondary or alternate thermal paths which would reduce or impair the accuracy of measurements to an appreciable extent. Thus, the most simple expedient consists in making use of a solid cylindrical radiation-absorbing body comprising a first thermocouple located axially and a second thermocouple located axially internally of a greater mass or at the periphery thereof. Since heat flows radially, the temperature difference measured between these thermocouples is proportional to the square of the radius of the transverse section of said solid cylinder and to the production of heat within the corresponding mass and is inversely proportional to the thermal conductivity of the body. However, the constructional design just mentioned cannot be adopted in practices since the measured temperature difference Delta t is not large enough to permit a really significant measurement within the dimensional limits permitted in the case of an arrangement within the bundle of pins of a fuel assembly, especially in a nuclear reactor of the pressurized-water type. Other design solutions derived from the preceding can also be contemplated with a view to guiding the flow of radial heat in a more effective manner. This is achieved by making provision within the radiation-absorbing body for a cavity in which are formed one or a number of insulating chambers between the central region and the peripheral region, these regions being connected to each other by means of at least one radial arm. While they do have the effect of increasing the measured temperature difference as a result of the higher resistance to heat transfer which is due to the present of the aforementioned insulating chamber, these design solutions still fall short of the standard of efficiency required to justify their adoption in a pressurized-water reactor and even in a fast reactor.
This invention relates to an improvement in the constructional arrangements recalled in the foregoing and accordingly makes it possible to obtain a measurement corresponding to an appreciable temperature difference. Furthermore all the objectives which have been stated in the foregoing are achieved by means of the device under consideration.